ML20055G544
ML20055G544 | |
Person / Time | |
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Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
Issue date: | 07/19/1990 |
From: | Office of Nuclear Reactor Regulation |
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ML20055G539 | List: |
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NUDOCS 9007230298 | |
Download: ML20055G544 (12) | |
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UNITED STATES l
NUCLEAR REGULATORY COMMISSION
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"%.....l SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.12g TO FACILITY OPERATING LICENSE NO. DPR-61 CONNECTICUT YANKEE ATOMIC POWER COMPANY f
HADDAM NECK PLANT DOCKET NO. 50-213 e
PART 1
, Reviews the licensee's request to implement an integrated surveillance program and corresponding TS changes.
PART 2 - Reviews the. licensee's request to change the TS to support the reactor vessel thermal shield removal and the reviews required by r
PART 3 - Reviews the portions of the submittal to support the thermal shield removal and associated TS changes related to the pressure vessel neutron irradiation surveillance programs.
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PDR ADOCK 05000213 p
PART 1 0F SAFETY EVALUATION RELATED TO AMEN 0 MENT NO.128
1.0 INTRODUCTION
By letter dated F e ruary 16, 1990, Connecticut Yankee Power Company (the licensee) provided information on the thermal shield removal program at hddam Neck.
Because the reactor vessel material surveillance capsules would be removed with the thermal shield.
The licensee proposed to utilize Millstone Unit 3 as a host reactor for a Haddam Neck capsule.
The licensee also proposed to revise the Technical Specifications to reflect the change in the surveillance program and to relocate the sut eillance schedule to the Updated Final Safety Analysis Report (UFSAR).
By letters dated March 29, and June 6, 1990, the licensee provided additional information relating to the integrated surveillance program according to Appendix H to 10 CFR Part 50.
2.0 EVALUATION 2.1 Integrated Surveillance Program The staff evaluated the licensee's proposed integrated surveillance program according to the four criteria in Appendix H to 10 CFR Part 50 and found the program acceptable.
The Haddam Neck cavity dosimetry program is discussed in Part 3 of the Safety Evaluation.
2.2 Capsule Withdrawal Schedule Tha licensee has vi*bdrawn four capsules.
Capsules A, F. H, and D were withdrawn at 2, 3, k, and 11 effective full power years gFPYs)2 respectively, at neutron fluences of 0.239, 0.471, 1.58, and 2.22 x 10 n/cm,
-respectively.
Appendix H to 10 CFR Part 50 requires reactor vessels to have surveillance programs meeting the requirements of ASTM E 185.
Because the predicted transition temperature shift at the vessel inside surface is approximately 150 F at end of life (E0L), E 185 recommends a minimum of four capsules to be withdrawn according to a certain schedule.
The four capsules withdrawn from Haddam Neck were withdrawn over a period approximately equivalent to that recommended for the first three capsules in E 185.
The E 185 recommended withdrawal schedule for the last capsule is E0L.
The licensee proposed to insert Capsule E from Haddam Neck into Millstone Unit 3.
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~2-i The licensee proposed to withdraw Capsule E at 32 EFPY, i.e., EOL for Haddam Neck.
The licensee proposed to store the remaining three Haddam Neck capsules j
in the spent fuel pool, Because Capsule E contains the weld metal which may
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be the limiting material with continued plant operation, the staff finds it i
appropriate to irradiate Capsule E at a host reactor.
Because the proposed withdrawal schedule for Capsule E is consistent with the recommendation of E 185, the staff finds the 11:ensee's request acceptable.
2.3 Technical Specification Changes The licensee also proposed to revise the Technical Specifications to reflect the change in the surveillance program and to relocate the surveillance L
schedule to the UFSAR.
The staff finds this request acceptable because Surveillance Requirement 4.4.1.1.2 in the Haddam Neck Technical Specifications requires compliance with Appendix H to 10 CFR Part 50 which in term requires a surveillance program.
The proposed changes would not alter the licensee's i
commitment to a surveillance program.
3.0
SUMMARY
As a result of the removal of the Haddam Neck reactor vessel thermal shield, the licensee requested to implement an integrated surveillance program in accordance with Appendix H to 10 CFR Part 50.
The licensee also requested corresponding Technical Specification changes. The staff has evaluated the licensee's. requests and found them acceptable.
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PART 2 0F SAFETY EVALUATION RELATED TO AMEN 0 MENT NO.128 l
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1.0 INTRODUCTION
By letter dated February 16, 1990, the Connecticut Yankee Atomic Power Company (the licensee) proposed changes to Haddam Neck Technical Specifications, J
section 3.4, Pressure-Temperature Limits. The proposed changes were submitted to support the reactor vessel thermal shield removal during the current (Cycle
- 15) refueling outage.
The submittal changed the effectiveness of the current P/T limits from 22 to 18 effective full power years (EFPY).
However, by letter dated March 29, 1990, the licensee reverted the effectiveness of the P/T limits from proposed 18 EFPY back to original 22 EFPY because the licensee judged that the n,otron fluence used to calculate the P/T limits for 18 EFPY was sufficiently high for the current P/T limits.
The current P/T limits were developed based on the data from actual surveillance capsules.
The P/T limits provide i
up-to-date P/T limits for the operation of the reactor coolant system during heatup, cooldown, criticality, and hyrirotest.
To evaluate the P/T limits, the staff uses the following NRC regulations and guidance:
Appendices G and H of 10 CFR Part 50; the ASTM Standards and the 4
ASME Code, which are referenced in Appendices G and H; 10 CFR 50.36(c)(2);
Regulatory Guide (RG) 1.99, Revision 2; Standard Review Plan (SRP) Section 5.3.2; and Generic Letter 88-11.
Each licensee authorized t? operate a nuclear power reactor is required by 10 CFR 50.36_to provide Technical Specifications for the operation of the plant.
In particular,10 CFR 50.36(c)(2) rew, es that limiting conditions of operation be included in the Technical Specifications.
The P/T limits are among the limiting conditions of operation in the Technical Specifications for all commercial nuclear plants in the U.S.
Appendices G.and H of 10 CFR Part 50 describe specific requirements for fracture toughness and reactor vessel haterial surveillance that must be considered in setting P/T limits.
An acceptable method for constructing the P/T limits is described in SRP Section 5.3.2.
Appendix G of 10 CFR Part 50 specifies fracture toughness and testing requirements for reactor vessel materials in accordance with the ASME Code l
and, in particular, that the beltline materials in the surveillance capsules I
be tested in accordance with Appendix H of 10 CFR Part 50.
Appendix H, in l
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These tests define the extent of vessel embrittlement at the-time of capsule withdrawal in terms of the increase in reference temperature.
Appendix G also requires the licensee to predict the effects of neutron irradiation on vessel embrittlement by calculating the adjusted reference temperature (ART) and Charpy upper shelf energy (USE).
Generic letter 88-11 requested that licensees and permittees use the methods in RG 1.99, Rev. 2, to predict the effect of neutron irradiation on reactor vessel materials.
This guide defines the ART'as the sum of unirradiated reference temperature, the increase in reference temperature resulting from neutron irradiation, and a margin to account for uncertainties in the prediction method.
Appendix H of 10 CFR Part 50 requires the licensee to establish a surveillance program to periodically withdraw surveillance capsules from the reactor vessel.
Appendix H refers to the ASTM Standards which, in turn, require that the capsules be installed in the vessel before startup and that they contain test specimens made from plate, weld, and heat-affected-zone (HAZ) materials of the reactor beltline.
2.0 EVALUATION The staff evaluated the effect of neutron irradiation embrittlement on each beltline material in the Haddam Neck reactor vessel.
The amount of irradiation embrittlement was calculated in accordance with RG 1.99, Rev. 2.
The staff determined that the two limiting materials a, 22 EFPY were the longitudinal weld (ARCOS B5 flux) with 0.22% copper (Ct), 0.10% nickel (Ni),
and an initial RT of -56'F;and the nozzle shell course plate W9807-6 with 0.11%Cu,0.20%N9dtand an initial RT f 34 F.
ndt The licensee has removed four surveillance capsules from Haddam Neck.
The results from capsules A and F were published in Battelle Reports BMI-1070 and BMI-0372.
The results from capsules H and D were published in Westinghouse reports WCAP-9339 and WCAP-10236.
The surveillance capsules contained Ch epy impact specimens and tensile specimens made from base metal, weld n ts1, and HAZ metal.
Forghelongitudinalweld,thestaffcalculatedtheARTtobe129.0'Fand 92.5 F at the 1/4T and 3/4T locations, respectively (T= reactor vessel beltline thickness).
Forthenozzleshellplage,W9807-6,thestaff calculatedtheARTtobe134' Fat 1/4Tand100[at3/4T.
Thestgffuseda revised surface neutron fluence of 4.77E19 n/cm and 2.53E19 n/cm for the weld and the plate, respectively. The neutron fluences were revised assuming the thermal shield is removed (Reference 4).
Although the ART of the weld is less than_the plate, the weld showed a higher RT shif t than the plate.
Also,theweldmayhaveahigherARTthanthepladtin the future because it is in a much higher neutron fluence zone and its copper and nickel contents are higher than the plate.
The licensee used the method in RG 1.99, Rev. 2, to calculate an ART value at 1/4T of 134.2 F and 101.0'F at 3/4T for nozzle shell course plate W9807-6.
Substituting the ART of 134.2 F into equations in SRP 5.3.2, the staff
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verified that the current P/T limits for hu tup, cooldown, and hydrotest meet the beltline material requirements in Appendix G of 10 CFR Part 50.
In addition to beltline materials, AppendiA G of 10 CFR Part 50 also imposes t
P/T limits based on the reference temperature for the reactor vessel closure flange materials.
Section IV.2 of Appendix G states that when the pressure exceeds 20% of the preservice system hydrostatic test pressure, the 1
tempe ature of the closure flange regions highly stressed by the bolt preload i
must exceed the reference temperature of the material in those regions by at least 120*F for normal operation and by 90*F for hydrostatic pressure tests and leak tests.
Based on the flange reference temperature of 10*F, the staff has determined that the proposed P/T limits satisfy Section IV.2 of Appendix G.
Section IV.B of Appendix G requires that the predicted Charpy USE at end of life (EOL) be above 50 f t-lb.
Nozzle shell plates, W9807-1, W9807-6, and W9807-8, have the lowest unirradiated USE in the transverse (weak) direction among all beltline materials. These are A302, Grade B plates. The licensee predicted that these plates will be above 50 ft-lb at end of life; however, the staff has a reservation about the licensee's prediction. The staff's reservation is based on 1) the licensee used a factor of 0.775 to convert the unirradiated USE of the plates from the lonoitudinal direction to the transverse direction whereas SRP 5.3.2 recsmmends a lower factor of 0.65, 2) the licensee applied a reduction factor of 12% to the unirradiated USE to predict the end of life USE for the plates; however, irradiated surveillance s)ecimer.s of the plates that have similar chemistry contents as the nozzle siell plates showed an USE reduction of as much as 20%, and 3) the removal of the thermal shield presents an uncertainty in the future increase of the l
neutron fluence.
The staff agrees with the licensee that the USE of the nozzle shell course plates requires further investigation. The staff will need more data to estimate whether the plates will be above 50 ft-lb at end of life.
3.0
SUMMARY
The staff concludes that the current P/T limits for the reactor coolant system for heatu), cooldown, leak test, and criticality are valid through 22 EFPY because tie limits conform to the requirements of Appendices G and H of 10 CFR Part 50. The licensee's submittal also satisfies Generic Letter 88-11 i
l because the licensee used the method in RG 1.99, Rev. 2 to calculate the ART.
l Hence, the current P/T limits may he maintained in the Haddam Neck Technical Specifications.
The staff judges that the low upper shelf energy of the nozzle shell course plates is an ongoing issue and not a stcrtup issue. The upper shelf energy of the plates is above the 50 ft-lb limit presently and will be above this limit L
for sometime. However, the staff recommends that the licensee follow the progress of B&W and CE Owners Groups studies on this issue because the licensee will have to address the staff's concerns mentioned above in the near i
future.
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4.0 REFERENCES
1.
Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessel I
Materials, Revision 2, May 1988 2.
NUREG-0800, Standard Review Plan, Section 5.3.2:
Pressure-Temper ~ure Limits i
3.
February 16, 1990 Letter from E. J. Mroczka (NU) to USNRC Document Control Desk, Suuject:
Haddam Neck Plant, Reactor Vessel Material Surveillance Program, Proposed Changes-to Technical Specification 4.
March 29, 1990, Letter from E. J. Mroczka (NU) to USNRC Document Control Desk,
Subject:
Haddam Neck Plant, Additional Information:
Reactor Vessel Material Surveillance Program S.
D. R. Ireland and V. G. Scotti, " Examination and Evaluation of Capsule A for the Connecticut Yankee Reactor Pressure-Vessel Surveillance Program," Battelle Memorial Institute Report BMI-1070, October 30, 1970 6.
J. S. Perrin et al, " Examination and Evaluation of Capsule F for the Connecticut Yankee Reactor Pressure-Vessel Surveillance Program,"
Battelle Memorial Institute Report BMI-0372, March 30,1972 7.
P. J. Fields and S, L. Anderson, " Analysis of Capsule H from the Connecticut Yankee Reactor Vessel Radiation Surveillance Program," WCAP-9339, September 1978 8.
S. E. Yanichko, et al., " Analysis of Capsule D from the Connecticut Yankee Reactor Vessel Radiation Surveillance Program," WCAP-10236, January 1983 9.
June 6, 1990, Letter from E. J. Mroczka (NU) to USNRC Document Control Desk,
Subject:
Haddam Neck Plant, Additional Information:
Reactor Vessel Material Surveillance Program 1
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PART 3 0F SAFETY EVAL.UATION RELATED TO AMENDMENT NO.128 j
1.0 INTRODUCTION
1 By letter dated February 16, 1990, the Connecticut Yankee Atomic Power Company (CYAPCO) the licensee for the Haddam Neck Plant submitted information to support the reactor vessel thermal shield removal and the associated technical l
specification changes to enable reactor operation after thermal shield removal J
(Ref. 1).
Additional information was submitted on March 29 and May 9, 1990 j
(Ref. 2 and 3).
The proposed technical specification changes are related to the pressure vessel neutron surveillance to satisfy the requirements of Appendix H, Appendix G and 10 CFR 50.62.
The thermal shield which was removed, provided surveillance capsule support.
CYAPCO decided not to reinsert the i
remaining four surveillance capsules into Haddam Neck.
The material irradiation and the surveillance roles of the capsules will be continued by (a) reinserting one of the capsules in Millstone 3 and (b) establishing a reactor cavity neutron dosimetry surveillance program.
The objective of both l
programs is to demonstrate reactor vessel operability for 32 effective full power reactor ysars, i.e., equivalent to 40 calendar years of operations.
2.0 EVALUATION 2.1 Thermal Shield Support Damage 4
The Haddam Neck thermal shield is 4.2 inch (10.67 cm) thick stainless steel cylinder, more than 38 tons in weight, supported as an inverted pendulum from the lower end of the core support barrel.
There were six support blocks at the bottom of the thermal shield, four displacement limiters and (initially) six flexures at the top of the shield.
All six flexures failed very early in the plant's life.
During the second in-service inspection (1987) extensive l
deterioration in the lower supports was found and wear in the upper displace-ment limiters.
The cause of the failure was attributed to flow induced thermal shield vibrations.
The licensee considered the alternatives and chose to remove the thermal shield.
The NRC is reviewing the removal of the thermal shield because of the following:
increased reactor vessel fluence, affecting:
plant life extension, pressurized thermal shock and low temperature operation, and impacts on the existing pressure vessel neutron surveillance program 2.2 Pressure Vessel Neutron Fluence The substitution of 10.67 cm of stainless steel with an equal thickness of water will cause an increase (of about 30 to 50%) of the E 1 1.0 MeV neutron flux to.the inside surface of the pressure vessel.
This is primarily due to the removal of the steel inelastic scattering at energies between 1.0 to 3.0 MeV when the steel is replaced with an equal thickness of water.
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These estimates were carried out using acceptable methods, i.e., a bencht3rked code ENDF/B-IV based cross-sections (SAILOR cross section set, Ref 4) and acceptable approximations, i.e., a P scattering and e S8 quadrature. The 3
results for the fast neutron flux on the inside surfaie of'the pressure vessel were correspondingly higher.
2.3 RT at 22 and 32 EFPYs PTS The new fluence estimate was used to calculate new RT values at 22 effective full power years (EFPY) (which is the preseESvalue in technical specifications 3/4 4-41 to 4-43 and B3/4 4-9) and 32 EFPYs, which corresponds to the end of the current license. The revised fluence values were combined I
with the methodology in Reg. Guide 1.99, Rev. 2 and the net effect was lower values for RT the new correN. The effect of the increased fluence was more than offset by ion in Reg. Guide 1.99, Rev. 2.
In addition the new values of RT criteb.have a very large mergin compared to the 10 CFR 50.62 screening The proposed technical specification changes are limited to the referencesforupdatingmaterialpropertiesasrequiredbyAppendixH(due to the lack of surveillance capsules). There are two such references in specification 4.4.9.1.2 and in the bases of 3/4.4.9.
2.4 The Integrated Surveillance Program AppendixH.SectionII.Cprovidesforan.integratedsurveillanceprogram(ISP) and includes four criteria which such programs must satisfy. The licensee addressed these criteria acceptably. The ISP allows for the irradiation of a surveillance capsule in a host reactor.
The licensee has chosen Millstone 3 as the host plant.
Licensee estimates indicate that the surveillance capsule will receive a conservative irradiation exposure at Millstone 3.
The licensee defined the removal schedule of this capsule at 25 EFPYs (equivalent to 31.25 calendar years) of exposure. This capsule is intended for reactor operation from 40 calendar years (current license) to 60 calendar years (extendedlicense).
However, as far as the surveillance function is j
concerned we note:- (a) that a core cavit established (see Section 2.5 below), (b) y dosimetry program will be that the pressure vessel has a very large estimated RT margin for the projected 20 calendar years of operation, and(c)thelicenshThas not yet decided whether they will continue irradiation of the three contingency capsules.
2.5 Cavity Dosimetry To replace the discontinued neutron dosimetry function due to the removal of the surveillance capsule the licensee will institute a cavity dosimetry program during this outage.
Because, there exists no generic approval of any cavity
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' dosimetry program, the licensee should provide sufficient information to justify the merits of their cavity dosimetry.
In view of the large margin of i
reactor vessel RT and given that this is a long term issue, the staff does not requir %edNke implementation of the cavity dosimetry program, however, we requirrjie the licensee submit the program for staff review prior to the next refue L.y outage.
2.6 Effect of Thermal Shield Removal on LOCA and Non-LOCA_ Transients Thermal shield removal will cause a slight increase in the reactor coolant flow and slight perturbations in the accident analyses.
The licensee dis-cussion showed that the effects are either conservative or insignificant.
Similarly r.; significant effect is expected from the increased gamma heating of the pressure vessel.
This conclusion is also supported by experience from previous instances of thermal shield removals.
3.0 TECHNICAL SPECIFICATION CHANGES
.The proposed technical specifications reflect the loss of the neutron dosimetry function of the. surveillance capsules. In particular:
Table 4.4-5 is deleted and the two references made to that table in 4.4,9.1.2 and Bases 3/4.4.9 are also deleted. We find the removal of the surveillance capsules acceptable as discussed in the evaluation, thus, the proposed technical specification changes are also acceptable.
4.0
SUMMARY
We have reviewed the CYAPC0 submittals related to the thermal shield removal issues at Haddam Neck.
Wc find that:
(a) There will be an 'ncrease in the E > 1.0 MeV neutron flux on the inside surface of the pressure vessel.
However, the estimated RT values usingReg. Guide 1.99Rev.2at32EFPYswillbelowerthaIkhe I
corresponding values with the thermal shield and Reg. Guide 1.99 Rev. 1.
We find the methodology and the results acceptable.
(b) CYAPC0 will institute an integrated surveillance program by continuing irradiation of one capsule at the Millstone 3 reactor.
This program satisfies the requirements of Appendix H and is acceptable.
(c) CYAPC0 will institute a core cavity dosimetry program which has not as of yet been specified.
The staff requires that the licensee commit to submit for staff review the cavity dosimetry program prior to the next refueling outage.
(d) The requested technical specification changes reflect the loss of the neutron dosimetry function of the surveillance capsules and are acceptable.
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5.0 REFERENCES
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Letter from E. J. Hroczka, CYAPCO to USNRC, "Haddam Neck Plant, Reactor Vessel Material Surveillance Program Proposed Changes to Technical Specifications," dated February 16, 1990.
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Letter from E. J. Mroczka, CYAPC0 to USNRC, " Additional Information Reactor Vessel Material Surveillance Program," dated March 29, 1990.
3.
Letter from E. J. Mroczka, CYAPCO to USNRC, "Haddam Neck Plant, Additional Information, Thermal Shield Removal Project," and attached WCAP-1250, " Connecticut Yankee Thermal Shield Removal Licensing Report,"
dated May 9, 1990.
4 SAILOR RSIC, DLC-76, " Coupled Self-Shielded 47 Neutron, 20 Gamma-Ray, P,
3 Cross Section Library for Light Water Reactor Applications."
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ENVIRONMENTAL CONSIDERATION Pursuant to 10 CFR 51.21, 51.32, and 51.35, an environmental assessment and finding of no significant impact have been prepared and published in the Federal Register on July 18, 1990 (55 FR 29283). Accordingly, based upon tho environmental assessment we have determined that the issuance of the amendment will not have a significant effect on the quality of the human environment.
CONCLUSION We have concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Dated: July 19, 1990 Principal Contributor:
S. Lee J. Tsua L. Louis l
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